IR 05000443/1989006

Insp Rept 50-443/89-06 on 890527-0630.No Deficiencies Found in Operation of Central Alarm Station.Major Areas Inspected: Plant Operation,Security,Operational Safety & Maint & Surveillance

ML20247Q786
Person / Time
Site:	Seabrook
Issue date:	07/27/1989
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20247Q746	List: IR 05000443/1989006 ML20247Q743
References
50-443-89-06, 50-443-89-6, NUDOCS 8908070247
Download: ML20247Q786 (12)

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION'I Report No.:

50-443/89-06

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• Docket No.:-

50-443 License No.: NPF-67.

Licensee:

'Public Service Company,of New Hampshire

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1000 Elm Street P

Manchester, New Hampshire 03105

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Facility:

-Seabroot Station, Unit No. 1 c.

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(Seabrook,. New Hampshire

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. Dates:

May 27. June.30,:1989 h>

Inspectors:

N. F. Dudley, Senior. Resident Inspector, Seabrook--Station

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M. Markley, Resident Inspector - Yankee Nuclear Power Station

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.R.'.Nimitz,' Senior Radiation Specialist, Facilities Radiation-Protection Section Approved By:

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7/27[oM ex D. R. Haverkamp,-Chief, Reactor P jects Jection 3B Date Inspection' Summary a.

Areas-Inspected Routine inspection conducted by resident and regional inspectors. Areas of inspection included plant operations, operational safety, maintenance

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and surveillance.

The inspections of plant operations in the areas of pre-paration for initial criticality, conduct of the startup test program and response to the unplanned reactor trip on June 22, 1989, are documented

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in NRC Inspection Reports 50-443/89-80, 89-81 and 89-82, respectively.

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Results:

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Security: The licensee effectively dealt with a large demonstration which included civil disobedience (paragraph 2.4).

No deficiencies were found in~the operation of the Central Alarm Station or the armory (paragraph 5).

Plant Operations: During the inspection period the plant was brought to normal operating temperature and pressure, the reactor was brought cri-tical for the first time, startup physics testing was successfully com-I pleted and an unplanned manual reactor trip occurred (paragraph 6.7) that prompted the NRC to dispatch an augmented inspection team to the site.

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l With the exception of the unplanned reactor trip, there were no opera-tional events during the preparation for, entry to and operation in operational mode 2.

Operator performance was considered by the inspector to be acceptable.

(As stated above, additional inspection: findings for events that took place during this period are documented in separate NRC reports.)

Maintenance and Surveillance: The modification and successful testing of the emergency feedwater pump turbine steam supply valves were observed and reviewed by the inspector (paragraph 3.1 and 6.3).

The successful boron analysis test of the post accident sampling system was reviewed by the inspector (paragraph 3.2). _The inspector observed adequate performance of selected maintenance and surveillance activities (paragraph 7).

Radiological Control: Two individuals _were slightly contaminated from a leaking check source (par,3raph 8.2).

The high alarm set point for a radi-ation monitor was incorrectly set during a release from a liquid waste test tank (paragraph 6.2).

Two open items concerning high radiation area

- controls and air sample locations were closed (paragraph 8.1).

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TABLE OF CONTENTS PAGE 1.

Persons Contacted.................................

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2.

. Summary of Facility and NRC Activities...............................

3.

Status of Previous Inspection Findings (IP 71707)*...................

3.1 Operability of EFW Steam Supply Va1ves..........................

3.2 Operability of Post Accident Sampling System....................

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Operational Safety (IP 71710)........................................

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Security (IP 71707)..................................................

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' Plant Operations'(IP 71707,93702)...................................

4 6.1 Power Operated Relief Valve Operability.........................

6.2 Rel ea s e of Li q uid Wa ste Ta n k....................................

6.3 Modification and Testing of EFW Turbine Driven Pump.............

6. 4 Inadvertent Engineered Safety Feature Actuation.................

6.5 Initial Criticality.............................................

6.6 Failure of Postmortem Report to Actuate.........................

6.7 Manual Reactor Trip During Natural Circulation.................

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Maintenance / Surveillance (IP 61726, 62703)..................

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Radiological Controls (IP 71707).....................................

8.1 Licensee Action on Previously Identified Items..................

8.2 Personnel Contamination..........................................

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Review of Periodic and Special Reports (IP 71707)....................

10. Management Meetings (IP 30703).....................................

• The NRC Inspection Manual Inspection Procedure (IP) that was used as inspection guidance is listed for each applicable report section.

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a DETAILS 1.

Persons Contacted W. A. DiProfio, Assistant Station Manager

• T. C. Feigenbaum, Vice President, Engineering, Licensing and Quality Programs

• D.'E. Moody, Station Manager

• J. E. Peschel, Regulatory Services Manager

• N. A. Pillsbury, Independent Review Team Manager

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G. S. Thomas, Vice President, Nuclear Production

• J. M. Vargas, Manager of Engineering

• J. J. Warnock, Nuclear Quality Manager

• Attended exit meeting conducted July 14, 1989.

Interviews and discussions with other members of licensee and contractor management and with their staffs were also conducted relative to the in-spection of items documented in this report.

2.

Summary of Facility and NRC Activities 2.1 Resident Inspector Activities One full time senior resident inspector was assigned to the site during the entire inspection period. The inspection inciuded 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> during back shift periods and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during deep backshift periods. Deep backshift inspection was conducted from 8:00 a.m. to 4:00 p.m. on May 27, 28 and 29, and on June 10, 11, 17, 18, 24 and 25, 1989.

In addition, the senior resident inspector participated in the team inspections described in paragraph 2.2.

2.2 Visiting Inspector and Management Activiti,es An NRC Region I reactor projects section ci.ief led a readiness as-sessment team inspection from May 27 through June 1, 1989, which will be documented in NRC Inspection Report No. 50-443/89-80.

The team evaluated licensee readiness for initial criticality through review of licensee records and direct inspection including continuous (twenty-four hour) shift observations during plant heatup.

A startup team performed continuous (twenty-four hour) test obser-vations of low power testing between June 14 and 22,1989, which will be documented in NRC Inspection Report No. 50-442/89-81. An NRC Region I inspector conducted an evaluation of the licensee's engi-neering and technical support function between June 19 and 23,1989, which will be docurented in NRC Insnection Report 50-443/89-07. The

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NRC Region I Deputy Regional Administrator teured the facility June 21 and 22, 1989 and conducted discussions with senior corporate man-agers.

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L The NRC Region I pressurized water recctor section chief led an aug-mented inspection team from June 28 through June 30, 1989, which will be~ documented in NRC Inspection Report No. 50-443/89-82. The team reviewed the causes, conditions and circumstances relevant to the manual reactor trip which occurred on June 22,.1989.

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2i3 Plant Status A low power. license, NPF-67, was issued to the licensee on May 26, 1989. ' Initial criticality.was attained-at 5:23 p.m. on June 13, 1989 and the low power' test. program was initiated. An unplanned manual reactor trip from 3% of rated power occurred during the natural cir-culation test on June 22, 1989 and the plant was cooled down to below 200 degrees F in operational mode 5 on June 29, 1989.

2.4 Public Demonstration On June 4,'1989, demonstrators scaled the owner controlled area

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In accordance with an agreement reached between state and local law enforcement agencies, the New Hampshire Yankee security department and the leaders of the demonstrations, approximately 680 demonstrators were stopped inside the fences by state police officers and arrested.

The demonstration was peaceful, no injuries occurred and no attempt was made by the demonstrators to enter the protected area.

3.

Status of Previous Inspection Findings 3.1 Operability of Emergency Feedwater Pump Turbine Steam Supply Valves (Closed) Construction Deficiency Report (CDR 86-00-05). While per-forming hot functional testing in late 1985, the licensee determined that the emergency feedwater (EFW) system was not delivering the re-quired minimum flow to the four steam generators. The licensee in-itiated design modifications which were implemented and the EFW.sys-e tem was retested in March 1987. ihe system met all test acceptance criteria, however several pump turoine overspeed trips occurred.

The NRC review of these tests are documented in NRC Inspection Reports 50-443/87-02 and 50-443/87-11 and as a result of NRC concerns, the licensee committed to conduct additional tests.

The licensee per-formed modifications to the steam supply valves (see paragraph 6.3).

2ad performed Turbine Driven Emergency Feedwater Start Verification Test STP-101 on June 11, 1989. The inspector reviewed the results of the test and verified that all acceptance criteria had been met and that steam header temperature was below 200 degrees F for two of the pump starts. The construction deficiency report is considered closed.

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3.2 Operability of Post Accident Sampling System (0 pen) Post accident sample system (PASS) operability:

In accordance with NUREG 0737 Task II.B.3 the sampling capability of the PASS is required to be demonstrated. On June 3,1989, the Acceptance Test -

Post Accident Sample Panel - RHR, CN86-1-22, rev. 3, was completed when tM plant was at normal operating temperature and pressure.

The acc 9tance test was performed to determine the accuracy of diluted PASS samples for boron and radionuclides. The inspector reviewed the completed procedure and found that the boron analysis was + 3.8% of the actual boron concentration which met the acceptance criteria of +

or - 5%. No radionuclides were detected due to the low level of radionuclides in the reactor coolant and the 1 to 250 dilution fac-tor. The licensee will attempt to measure the radionuclides again at a higher power level. A comprehensive NRC review of the operability of PASS will be performed.

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Operational Safety The inspector conducted a complete safety system walkdown of the safety injection and accumulator systems and verified the proper lineup for the residual heat. removal system from the main control room in accordance with requirements for operational mode 2.

No discrepancies were identified in the system lineups by the inspector.

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Security The inspector conducted a tour of the Central Alarm Station and the armory, and held discussions with security management on contingency plans for expected large demonstrations. The inspector was informed that ex-tensive planning had been conducted with local and state law enforcement agencies. Two unrelated threats had been made against the licensee. One through a TV news station and another through the local police. The treats were investigated, including notification of the FBI in one case,

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and were' deemed to be noncredible. The inspector noted no deficiencies in l

the operatier of site security during his tour and had no further ques-l tions.

6.

Plant Operations

6.1 Power Operated Relief Valves Operability I

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On June 1,1989 the control room operator questioned the operability

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pressure controller (RC-PK455) being in manual.

echnical Specifi-cation (TS) 3.4.4 requires both PORVs to be operable in Mode 3 or the associated block valve to be shut within one hour.

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I Through discussions with the licensee, the inspector determined that the master pressure controller provides the input signal to the train

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"A" PORV actuation logic circuit. When the master controller is in automatic. the pressure signal to the actuation circuit is provided by a pressurizer pressure instrument, however, when the master con-troller is in manual the pressure signal is the demanded pressure signal and precludes automatic operation of the PORV. The train "B" PORV actuation logic circuit receives a pressure signal directly from l

a pressurized pressure instrument.

The licensee prepared a technical clarification for TS 3.4.4 which determined that the train "A" PORV is operable when the master pres-sure controller is in manual based on the fact that no credit is taken in any' accident analysis for automatic PORV actuation, that no surveillance requirements exist for verifying automatic actuation and that the required action for an inoperable PORV conflicts with any l

presumed requirement for automatic operation. The inspector reviewed l

the justification contained in technical clarification to TS 3.4.4 I

and had no further questions on the operability of the train "A" PORV.

6.2 Release of Liquid Waste Tank l

During review of station information report 89-36, the inspector noted that on June 2,1989 the liquid waste test tank had liquid dis-l charged from it with the radiation monitor high alarm, which trips the discharge isolation valve, incorrectly set above the required trip set point. The alert alarm set point which provides an alarm prior to radiation levels reaching the high alarm set point was pro-perly set and did not alarm during the discharge. This indicates l

that radioactive effluent release limits were not exceeded.

During discussions with the licensee the inspector determined that the health physics (HP) technician had incorrectly entered the set point following the functional test of the waste discharge isolation valve and had not verified the set point. The Unit Shift Supervisor (USS) had verified the correct set point prior to the conduct of the waste discharge isolation valve functional test, which requires changing the set point, instead of after the final entry of the set point.

The licensee has revised the release permit to require the l

sign off of the verification of the set point after the completion of the functional test. The inspector raised concerns as to whether the discharge procedure was properly followed and what the generic im-plications are for other inter-organizational procedures which re-quire operation's verification.

This item is unresolved (UNR 50-443/89-06-01).

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4 6.3 Modification and Testing of Emergency Feedwater Turbine Driven Pump Background'

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~ During the emergency' feedwater (EFW) system testing performed during hot: functional testing in March L1987, conditions for cold starts of the EFW turbine driven.purap could not be established unless manual'

isolation ' valves were shut due to the significant leakage past the motor: operated isolation valves. LNew valve internals were installed in valves:FW-393 and FW-394,.the motor operated isolation valves fyr.

main steam line headers A and B, respectively. and FW-395, the motor operated. valve upstream of the'EFW pump: turbine trip valve. The new valve internals were designed to minimize steam leakage to..the EFW pump turbine during. normal operations and to open in sufficient time

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to meet the EFW operability criteria. The licensee committed to-demonstrate five starts.of the EFW system after the new valve in-ternals were installed.

Description of Testing and Design Modifications

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On June 5,1989 the steam turbine supply valves FW-393 and 'rW-394 failed to open during performance of special test procedure STP-101.

After discussions with the valve manufacturer, Hammel Dahl, the lic-ensee made modifications to the internals of FW-395 by increasing the number and size of the orifice vent holes to the valve pilot chacer in an attempt to' reduce the differential pressure across the valve.

With valves FW-393 and FW-394 open and the turbine trip valve shut, valve FW-395 was tested satisfactorily at various steam pressures.

However, when tested under full flow conditions, valve FW-395 failed to;open.

Further diagnostic tests revealed that the quick opening of valve FW-395 created flow instabilities which caused the turbine to trip on overspeed.

In consultation with Masoneilan-Dresser Industries, a valve manufacturer, the licensee developed additional changes to the valve internals of FW-393 and FW-394.

On June 10, 1989 the redesigned valve internals were installed in FW-393 and FW-394, and the internals used during hot functional tests in 1987 were installed in FW-395.

Special Test STP-101 was completed

satisfactorily on June 11, 1989.

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During performance'of diagnostic testing on June 7, 1989, the EFW g

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turbine-driven pump was lined up to recirculate water to the conden-sate storage tank. Based on test data, review by the technical sup-port department determinid that the turbine oversneed resulted in greater than 1600 psig in the recirculation line. The department requested an engineering evaluation (RES89-354) to determine the effects of the potential overpressurization. The inspector reviewed the completed engineering evaluation which concluded that the over-pressurization did not exceed any component design stress limits and had no further questions.

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Evaluation The inspector observed portions of the diagnostic testing of FW-395, the actuator modifications and the functional testing of the three l

steam valves. The inspector reviewed the approved working copies of l

some modification work packages and the completed special test pro-l cedure STP-101. The inspector discussed the details of the modifi-l cation and testing with the licensee.

The licensee conducted extensive diagnostic testing and design work during development and implementation ~of the final system and valve l_

modifications. The inspector found that the original design of the s

new valve internals for the EFW steam valves was unique for the ap-plication for which they were used by NHY. Additionally, the need for a stronger actuator was not recognized by the engineering de-partment at the time of the ir.itial installation of the new valve internals. The testing and modification of the valves were ade--

quately controlled with sufficient quality control and management involvement. Additional NRC review of'EFW valve performance will be documented in a separate NRC specialist inspection report.

6.4 Inadvertent Engineered Safety Feature Actuation

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On June 10, 1989 with the plant in operational Mode 3 and the con-tainment on-line purge (COP) system in operation, an inadvertent containment ventilation actuation signal was initiated due to a failed radiation monitoring detector. All systems functioned nor-i mally, the detector was replaced and the C0P system was returned to i

service. The appropriate emergency notification call was made based on the spurious engineered safeguards features actuation. The in-spector had no further questions.

6.5 Initial Criticality On June 13, 1989, at 5:23 p.m. following a planned boron dilution, the reactor achieved a self sustaining chain reaction. The approach to and achievement of initial criticality were conducted without error.

6.6 Failere of Postmortem Report to Actuate Automatically On June 17 and 18 the inspector observed two planned trips associated with low power testing.

The operators properly used the emergency operating procedures and correctly transitioned to normal operating procedures.

All equipment operated as designed with the exception of thd actuation of the postmortem report. The postmortem report is intended to actuate automatically on a reactor trip and to monitor

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and record plant parameters for post trip reviews.

The computer which produces the report : cans the various signals which cause a reactor trip, including the position of the manual trip switch, every

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five= seconds. During the first manual reactor trip, the manualItrip

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switen was not in,the trip position when scanned by the computer.

Therefore? the postmortem report was not automatically initiated but

.was. manually initiated one minute after theLtrip..Since'the trip was planned and occurred at a low power level, the lost minute of data was. not. necessary to reconstruct the plant transient.

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The licensee has modified the postmortem report actuation logic to1 include reactor trip breaker position which will. initiate the post mortem report whenever a reactor trip breaker opens. 'The' inspector had no'further questions.

6.7 Manual Reactor 1 Trip During Natural' Circulation Test On. June 22,'1385 during-a' natural circulation test, a condenser steam dump valve' failed open resulting in an unexpected cooldown'of the reactor coolant system.. At the time of the malfunction'the. reactor coolant pumps'were tripped and.the reactor was critical at approxi-

.mately~3%'of rated power.. The cooldown caused pressurizer level to

~ decrease below 17% of-full capacity which automatica11y'deenergized

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pressurizer heaters and isolated letdown. When the malfunctioning.

. steam dump was closed, the cooldown terminated and the pressurizer level; increased to about 22% of full capacity, which caused pres-

surizer pressure to also increase rapidly. The reactor was manually

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tripped. prior.to reaching the auto,aatic trip set point for high pres-sure, An NRC' augmented team inspection was initiated to evaluate this' event and'present their findings in'NRC inspection report 50-443/89-82.

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~ Maintenance / Surveillance On June 13, 1989 the inspector observed the corrective maintenance work.

performed in accordance with work request 89-002904 for investigation and repair of the digital rod positicn indication non urgent alarm. The.in-spector later reviewed the completed work requests and determined that the maintenance had been performed satisfactorily. The inspector had no con-cerns.

. On June 22, 1989 the inspector observed the solid state protection system Train B actuation logic test (1-SSPS-TRN-B) of the P7, reactor protection relay. The I&C technicians properly followed the procedure, maintained formal communications, recorded data and sought supervisory assistance y'

when necessary.

Licensee quality control also monitored the' surveillance.

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The inspector had no concerns.

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Radiological _Cpntrols 8.1 Licensee Action on Previously Identified Items (Closed) Inspector Follow Item 86-25-02: Control of Potentially Locked High Radiation Areas. The irspector's review of the licen-see's program found that the licensta has pre-identified those areas that will have radiation levels necessitating locking of the areas.

The licensee has established controls, either key and lock or key card, for access to those areas.

Station radiation levels will be monitored and areas will be locked prior

' exceeding technical specification locking requirements.

Some areas. such as the reactor containment, have already been locked. The implementation of the high radiation area access control program will be reviewed during future inspections.

Inspector follow item 86-25-02 is closed.

(Closed) Inspector Follow Item 86-39-11:

Selection of Routine Air Sample Locations. The inspector's revier of the licensee's program found that the licensee has evaluated the station locations where area air monitors should be placed. Areas have been pre-selected and a scheduled established for san,pling implementation. An adequate supply of instruments is available.

Inspector follow item 86-39-11 is closed.

8.2 Personnel Contamination On June 1,1989 an I&C technician alarmed a whole body frisker when exiting the radiological controlled area (RCA). The individual's shirt was found to be contaminated on the upper left chest with a speck of chlorine 36 (C1-36). The skin dose to the individual was approximately 300 millirem. The contamination appears to have ori-ginated from a leaking. check source which was associated with a radi-ation detector which had been removed from service and stored in a i

drawer in the I&C shop since 1986. One other individual was found to

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have sustained low level Cl-36 shoe contamination.

The licensee's corrective actions included surveying of the I&C lab and other administrative areas, confiscation of approximately

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twenty-four detectors and sources from the I&C shop and the warehouse

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and conducting discussions with the detector manufacturer to assess any Part 21 deportability issues resultirg from the event. The lic-ensee concluded that no Part 21 repnrt was necessary based on the wide dissemination of information regarding leaking sources within the industry.

The licensee's response to the event was extensive, well-focused and adequately addressed the specific problems. The

inspector had no further questions.

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Review of Periodic and Special Reports The inspector revieweed the following periodic and special reports and '

,sund no deficiencies or concerns.

10 CFR 50.59 Quarterly Report issued May 30, 1989

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Monthly Operating Report issued June 8, 198?

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10. Management Meetings At periodic intervals during this inspection, meetings were held with

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senior plant management to discuss the findings. A summary of the report period was also discussed at the conclusion of the inspection and prior to report issuance.

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